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05000263/FIN-2018003-0130 September 2018 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants; which requires, in part, that equipment qualified by test must be preconditioned by natural or artificial aging to its end of life or a shorter designated life considering all significant types of degradation which can have an effect on equipment function. Contrary to the above, on June 2, 2018, the licensee determined that EQ evaluation 608000000032, of MO2034, MO2035, MO2075, and MO2076 (HPCI and RCIC Steam Line Isolation Valves) internal actuator cables, failed to consider the temperature rise due to the high temperature process fluid in the vicinity of the affected components when aging (preconditioning) them and the unaccounted temperature rise shortened the life of some components to the point that they were no longer EQ qualified to the end of planned life. The unaccounted for process fluid temperature increases were verified by the licensee when thermography of the associated valves was performed. The licensee performed a prompt operability determination, entered the issue into the corrective action program (CAP) as CAP 501000012766 and performed a thermal life analysis engineering evaluation. Long-term corrective actions include replacement of the internal actuator cables during the next refueling outage. 10 Significance/Severity Level: This finding was more than minor because the performance deficiency was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, HPCI and RCIC Steam Line Isolation Valves are designed to provide reactor coolant pressure boundary, required for a safe reactor shutdown following a Design Basis Accident or transient. The finding was of very low safety significance (Green) because it was a design or qualification deficiency, did not involve an actual loss of safety system, did not represent actual loss of a safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hrs. Corrective Action Reference: 501000012766
05000263/FIN-2018012-0230 September 2018 23:59:59MonticelloNRC identifiedFailure to Implement Adequate Freeze Protection Monitoring for Condensate Storage Tank Instrumentation Piping in Response to Industry Operating ExperienceThe inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to ensure conditions adverse to quality are promptly identified and corrected. Specifically, the licensee failed to identify that monitoring of the CST instrument line heat tracing performed every 30 days was inadequate to assure the safety-related CST level instrumentation remained operable during extreme cold weather conditions
05000263/FIN-2018012-0130 September 2018 23:59:59MonticelloNRC identifiedInboard Main Steam Isolation Valve Closure Time Test Acceptance Criteria Did Not Account for the Design Basis Accident Containment Back Pressure and Pneumatic Supply Operating PressureThe inspectors identified a Green finding and an associated NCV of Title 10 of the Code of Federal Regulations(CFR), Part 50, Appendix B, Criterion XI, Test Control, for the failure to assure that applicable requirements and acceptance limits contained in the inboard main steam isolation valve (MSIV) design documents were incorporated into their test procedure. Specifically, the inboard MSIV closure time acceptance criteria contained in Functional Test Procedure 0255-07-IA-2, Main Steam Isolation Valve Functional Checks Test, did not account for the elevated containment pressure and the expected lower pneumatic supply pressure expected during design basis accidents.
05000263/FIN-2018012-0330 September 2018 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThis violation of very-low safety significance was identified by the licensee and has been entered into the licensee CAP. Therefore, this finding being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.Enforcement:Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Updated Final Safety Analysis Report, Appendix I,Evaluation of High Energy Line Breaks Outside Containment,Table I.5-2, Table of System Effects,Revision 36P, listed the Division II emergency power system as available during HELBs outside containment. Contrary to the above, on July 29, 1974, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically,the Division II emergency power system would not be available during a HELB outside containment.Procedure B.09.07-05, Operations Manual Section 4.16 kV Station Auxiliary, Revision 53,had actions that required entry into the lower 4kV area to permit repowering Division II emergency power systems but this area would be inaccessible during the event. Significance: The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences.Specifically, the performance deficiency resulted in a condition were the Division II emergency power system would not be available during HELBs outside containment. The inspectors assessed the significance of the finding using the SDP in accordance with IMC 0609, 11 Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 2, Mitigating System Screening Questions,and concluded the violation was of very-low safety or security significance (Green)because the licensee reasonably demonstrated an alternate strategy was available to timely reach and maintain cold shutdown conditions. Corrective Action References: CAP501000011837, CAP 50100001593
05000263/FIN-2018002-0130 June 2018 23:59:59MonticelloLicensee-identifiedLicensee-Identified Violation

This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section2.3.2 of the Enforcement Policy.Enforcement: Violation: Title 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements of 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(8) requires, in part, that a licensee must provide and maintain adequate emergency facilities and equipment to support the emergency response plan.Contrary to the above requirements, on March 23, 2018, the licensee identified the site failed to maintain the effectiveness of the emergency plan by not providing and/or maintaining equipment capable of measuring the Immediately Dangerous to Life and Health (IDLH) concentrations for several toxic chemicals as required to properly classify an Alert Emergency Action Level (EAL). Specifically, while performing an emergency equipment inventory, the licensee identified that detector tubes (Draeger tubes) available to measure chlorine gas concentrations were not capable of measuring the IDLH concentration of 10 ppm required to identify the threshold level for classifying an Alert EAL (HA 3.1) since the measurement range of the available sample tubes was 50500 ppm.The inability to properly classify the Alert EAL represented a Loss of Emergency Assessment Capability and resulted in the licensees submission of Event Notification Report # 53298 in accordance with the requirements of 10 CFR 50.72(b)(3)(xiii). An immediate extent of condition review performed by the licensee identified additional deficiencies in adequate sampling methods for determining IDLH concentrations for Butadiene, Ethylene Dichloride, and Gasoline. Additionally, the licensee identified that in April 2015 there was missed opportunity to correct this deficiency when an Emergency Preparedness (EP) Coordinator, performing a Control Room Emergency Equipment Inventory, identified the need to order and replace the existing chlorine detector tubes. The EP Coordinator added the incorrect detector tubes to the existing inventory form without validating the tubes detection range and accuracy to ensure it was capable of detecting the IDLH threshold concentration level of 10 ppm.Upon identification of the issue, the licensee implemented compensatory measures for determining the EAL classification and entered the issue into the corrective action program (CR 501000009876). On May 08, 2018, the licensee implemented the sites new EAL classification procedure that was developed using NEI 9901, Revision 6, which does not require atmospheric sampling (use of detection tubes) for classification of EAL HA 3.1.Significance/Severity Level: Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.81, the inspectors determined this finding was

10 of very low safety significance (Green) because a significant amount of equipment necessary to implement the E-plan was not available or functional to the extent that any key ERO member could not perform his/her assigned functions, in the absence of compensatory measures (Degraded Planning Standard), specifically the ability to accurately classify the Alert EAL. Determining the finding significance using IMC 0609, Appendix B, Table 5.41, results in the same finding significance (very low significance) since the performance deficiency would have rendered an EAL initiating condition ineffective such that the Alert would have been declared in a degraded manner.Corrective Action Reference: 501000009876, CR Toxic Gas Detector Tube.
05000263/FIN-2018001-0231 March 2018 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationViolation: Title 10 CFR 50.59(d)(1) requires, in part, that the licensee maintain records of changes to the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c).These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to Paragraph (c)(2) of this section.Title 10 CFR 50.59(c)(2)(ii) requires that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the Final Safety Analysis Report (FSAR) (as updated).Technical Specification (TS) 3.3.1.1, Reactor Protection System (RPS) Instrumentation, states the RPS instrumentation for each function in Table 3.3.1.11 shall be operable. As specified in Table 3.3.1.11, Function 5, Main Steam Isolation Valve (MSIV) - Closure (8 channels) and Function 8, Turbine Stop Valve (TSV) Closure (4 channels) are required to be operable in Mode 1. TS 3.3.1.1, Condition C.1 states with one or more functions with RPS trip capability not maintained, to restore RPS trip capability in 1 hour and was applicable to both the MSIV and TSV RPS logic functional testing.Contrary to the above, on March 7, 2009 and July 11, 2009, the licensee failed to perform and maintain a written evaluation as required by 10 CFR 50.59(d)(1) to demonstrate a change to its facility did not require a license amendment. Specifically, the licensee incorrectly concluded in its 10 CFR 50.59 evaluation SCR080319, dated September 29, 2008, that no license amendment was required prior to implementing two surveillance test procedures; 0009 Turbine Stop Valve Closure Scram Test Procedure, Revision 16 on March 7, 2009 and; 0008 Main Steam Line Isolation Valve Closure Scram Test Procedure, Revision 20 on July 11, 2009. The test fixture was applied during quarterly surveillance testing through September 16, 2017.Implementation of procedures 0008 and 0009, respectively, resulted in the loss of RPS trip Function 5 (MSIV) and Function 8 (TSV) by bypassing more than the TS minimum allowed inputs per channel to maintain functionality, thereby violating the requirements of TS 3.3.1.1. Loss of these functions resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated) as specified by 10 CFR 50.59(c)(2)(ii).On November 14, 2017, the licensee generated CAP 501000005391 after conducting an operating experience evaluation of a similar event at another station concluding the event was applicable to the Monticello Plant. The surveillance procedures were immediately quarantined and subsequently revised on December 8, 2017 and December 11, 2017, to remove the use of the RPS test fixture.Significance/Severity Level:Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined this finding was of very low safety significance (Green) because it did not affect a single RPS trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown.The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV The disposition of this violation closes LER 05000263/201700600.Corrective Action Reference: 501000005391
05000263/FIN-2018001-0131 March 2018 23:59:59MonticelloNRC identifiedFailure to Follow Procedure for Storage of Equipment Near Safety-Related EquipmentThe inspectors identified a finding of very low safety significance (Green) with an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B Criterion V for the failure to accomplish activities affecting quality as prescribed by documented procedures. Specifically, the licensee failed to follow procedure 4 AWI04.02.01, Housekeeping for storage of items or equipment near safety-related equipment. On two separate occasions, the inspectors identified items being stored near safety-related equipment that did not comply with procedure requirements.
05000263/FIN-2017409-0231 December 2017 23:59:59MonticelloLicensee-identifiedLicensee-Identified Violation
05000263/FIN-2017406-0131 December 2017 23:59:59MonticelloNRC identifiedSecurity
05000263/FIN-2017409-0131 December 2017 23:59:59MonticelloNRC identifiedSecurity
05000263/FIN-2017004-0131 December 2017 23:59:59MonticelloSelf-revealingFailure to Maintain Radiation Exposure ALARAA finding of very low safety significance (Green) was self-revealed due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees radiological work planning and work control program. Specifically, the licensee failed to properly incorporate ALARA strategies, insights while planning, and executing work activities during the 1R28 refueling outage. The Reactor Water Cleanup (RWCU) Inlet Outboard Isolation Valve MO2398 was scheduled for replacement during the outage. The initial dose estimate for this activity was 4.5 person-rem. However, 13.776 actual person-rem of dose was received. This issue was caused by poor radiological planning and work execution of this task. The licensee entered this issue into their Corrective Action Program (CAP) item 1558234. The finding was more than minor because it was associated with the program and process attribute of the Occupation Radiation Safety Cornerstone. Additionally, this issue affected the cornerstone objective of ensuring the adequate protection of the workers health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the finding is very similar to IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance (Green) because Monticello Nuclear Generating Plants current 3year rolling average collective is 64.637 person-rem (20142016). This is less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. This finding had a cross-cutting aspect in the area of Human Performance, related to the cross-cutting aspect of Work Management, in that the outage plan did not adequately plan, control and execute work activities to ensure the RWCU Inlet Outboard Isolation Valve MO2398 replacement remained ALARA. (H.5)
05000263/FIN-2017004-0231 December 2017 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix R, Section III.G.2.a, Fire Protection of Safe Shutdown Capability. Specifically, the licensee identified that the structural steel located in the plant administrative building (PAB) basement supporting the cable spreading room (CSR) floor did not have a 3hour fire rating as required by 10 CFR 50, Appendix R, Section III.G.2.a. Title 10 CFR 50, Appendix R, Section III.G.2.a, requires, in part, that where separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3hour rating is provided, structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier. Contrary to the above, since 1982, the licensee failed to protect the structural steel supporting the fire barrier between the cable spreading room and fire area IV. This failure was identified by the licensee on August 4, 2016 during an Appendix R self-assessment and addressed in CAP 1530637. The licensee issued LER 201600201 in response to this Appendix R non-compliance and implemented the immediate corrective action (compensatory measure) of an hourly fire watch in the affected fire area. The licensee conducted an exposed steel fire simulation and evaluation to understand the performance of the unprotected steel in the event of a fire in the PAB. The inspectors reviewed the licensees simulation and evaluation. This finding was determined to be more-than minor because the performance deficiency was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the structural steel supports the fire barrier between the CSR and the PAB basement and a failure to protect the structural steel from fire damage would degrade the fire barrier separating the CSR and PAB. This could result in a fire in the PAB spreading to the CSR due to the degraded fire barrier resulting in an evacuation of the control room. The only means for operators to shutdown the reactor using the ASDS panel would require travel through the PAB fire area where a fire event is occurring. Therefore, this finding impacted the safe shutdown capability of the plant. After review of the licensees exposed steel fire simulation and evaluation, the finding was determined to be of very low safety significance (Green) because the licensee demonstrated that the unprotected structural steel would provide at least one hour of fire endurance rating under a fire event in the PAB.
05000263/FIN-2017007-0130 September 2017 23:59:59MonticelloNRC identifiedInadequate Fire Barrier Inspection ProcedureThe inspectors identified a finding of very-low significance (Green) and an associated Non-Cited Violation of License Condition 2.C.4 of the Monticello Nuclear Generating Plant,Unit No. 1,Renewed Facility Operating License for implementing an alternative compensatory measure that was adverse to safety shutdown.Specifically, the licensee approved the installation of a temporary fuel oil pump, in lieu of a continuous fire watch, which reduced the defense in depth of the Fire Protection Program.The inspectors determined that the use of a temporary fuel oil pump in the event of afire, in lieu of a continuous fire watch, constituted an adverse change to the Fire Protection Program,was contrary to License Condition 2.C.4 and a performance deficiency. The performance deficiency was more-than-minor because it affected the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the use of the alternative compensatory measure reduced the defense in depth of the Fire Protection Program by failing to provide compensatory measures to reduce the likelihood of occurrence of a fire and failing to provide prompt detection of a fire.In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined the finding affected the Initiating Events cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process.The inspectors determined that the finding represented a low degradation and was screened as having very-low safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F,because repair activities were in place that would have maintained safe shutdown(SSD)conditions and were reasonably achievable.This finding had a cross-cutting aspect in the Conservative Bias component of the Human Performance cross-cutting area. Specifically, the licensee implemented an alternate compensatory measure that only focused on the emergency diesel generator operability and hence, the post-SSD strategy of the plant without considering the defense in depth requirements of their Fire Protection Program to prevent, detect, and suppress a fire that could affect equipment needed for SSD of the plant.
05000263/FIN-2017002-0230 June 2017 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of TS 3.7.1, Residual Heat Removal Service Water (RHRSW) System; which requires, in part, that two RHRSW subsystems shall be operable in Modes 1, 2, and 3 or per Condition A, One RHRSW subsystem inoperable; the RHRSW subsystem must be restored to OPERABLE status within 7 days or the applicable conditions and required actions of Limiting Condition forOperations 3.4.7, Residual Heat Removal Shutdown Cooling System Hot Shutdown, for RHR shutdown cooling made inoperable by RHRSW System must be entered. Contrary to the above, on March 27, 2017, the licensee exited the requirements in TS 3.7.1, with a Tag Section still hanging, rendering B RHRSW subsystem inoperable, while in Mode 1. This was identified by the licensee when the maintenance organization notified operations that work was complete, and the Tag Section was released. The licensee reentered TS 3.7.1, Condition A, entered the issue as CAP 1554105 and assigned a Human Performance Event Investigation. A crew clock reset was also taken as well as communicating lessons learned to the entire plant organization.This finding was more-than minor because the performance deficiency wasassociated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, RHRSW System is designed to provide cooling water for the RHR System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident or transient. Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post-accident heat loads, assuming the worst case single active failure occurs coincident with the loos of offsite power. The finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not involve an actual loss of safety system, did not represent actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-Tech Spec Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hours.
05000263/FIN-2017002-0130 June 2017 23:59:59MonticelloSelf-revealingLow Reactor Water Level During Shutdown of 11 Reactor Feedwater PumpA self-revealed finding of very-low safety significance and a Non-Cited Violationof Technical Specification 5.4.1.a occurred on April 15, 2017, due the licensees failure to establish, implement and maintain procedures regarding shutdown operations. Specifically, Operations Manual B.06.05-05 did not account for the state of the opposite train of feedwater when shutting down the 11 Reactor Feedwater Pump. Licensee use of the inadequate procedure placed equipment in a configuration where no condensate flow path to the reactor existed causing reactor water level to lower to a point where trip/isolation set-points were reached. This caused an unplanned Reactor Protection System (RPS) trip and Partial Group II Isolation. The licensee initiated Corrective Action Program (CAP) 1555785 to document the reactor water level transient, RPS trip and Partial Group II Isolation. Immediate corrective actions includedopening the 11 Reactor Feedwater Pump discharge valve to restore reactor water level allowing reset of the Group II isolation and RPS trip. Subsequent licensee actions included development of expectations via an Operations Memo and revision to Operations Manual B.06.0505 as well as Procedure 2204 and Procedure 2167 to ensure abnormal equipment lineups are addressed such that unexpected procedure interactions are avoided.The inspectors determined the failure to establish, implement and maintain procedures regarding shutdown operations as required by Technical Specification 5.4.1.a was a performance deficiency that required an evaluation. The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, and IMC 0609, Appendix A, Exhibit 1, Section B, and determined a detailed risk evaluation was required because the finding caused a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of feedwater). A Senior Reactor Analyst performed a detailed risk evaluation using bounding assumptions and the change in Core Damage Frequency was calculated to be 9E7/year (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Change Management aspect, because licensee leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000263/FIN-2017010-0130 June 2017 23:59:59MonticelloSelf-revealingFailure to Ensure Adequate Design Controls During Installation of Flexible Hose on High Pressure Coolant Injection Auxiliary Oil SystemGreen. A finding of very low safety significance and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, was self-revealed as a result of an equipment cause investigation following failure of a pipe nipple in the safety-related piping for the HPCI system on March 22, 2016. Specifically, during original installation of the HPCI system, the licensee failed to correctly install a flexible hose to isolate vibrations in the system. Immediate corrective actions taken by the licensee included installing the flexible hose in the correct location to ensure isolation of vibrations in the system and performing walkdowns of other risk-significant systems to verify flexible hoses were installed in accordance with design. The issue was captured in the licensees corrective action program under CAP 1516361. 3 The inspector determined that the failure of the licensee to implement adequate design control measures and assure any deviations from Design Drawing NX82928 were properly controlled during installation of the flex ible hose in the HPCI system was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The performance deficiency was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to install the flexible hose in the correct location in the HPCI oil system resulted in increased vibrations and loads throughout the HPCI system which had the potential to further degrade and decrease the reliability of the system. The finding was screened using Inspection Manual Chapter 0609, Appendix A, against the Mitigating Systems Cornerstone and determined to be of very low safety significance (Green), because the inspectors answered No to all of the questions in Exhibit 2, Mitigating Systems Screening Q uestions, Section A, Mitigating SSCs and Functionality. A cross-cutting aspect was not assigned to this finding since the performance deficiency occurred during the origi nal installation of the HPCI system and was determined not to be indicative of current licensee performance.
05000263/FIN-2016007-0131 December 2016 23:59:59MonticelloNRC identifiedInadequate Procedure for Identification of Significant Conditions Adverse to QualityThe inspectors identified a finding of very low safety significance and non-cited violation of Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to prescribe a procedure appropriate to the circumstances with respect to the identification of a significant condition adverse to quality (SCAQ). Specifically, FPPAARP01, CAP Action Request Process, provided an overly restrictive definition of what constituted a SCAQ. Consequently, the failure to provide an adequate definition of a SCAQ could result in a failure to identify a SCAQ and therefore, failure to implement corrective actions that preclude repetitive failures of safety-related equipment. The licensee entered this issue into the CAP as action request (AR) 1536735. The inspectors determined that the licensees failure to prescribe a procedure appropriate to the circumstances under FPPAARP01 was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because, if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Although, this issue could potentially affect each of the Reactor Safety Cornerstones, the inspectors elected to evaluate this issue under the Mitigating Systems Cornerstone because inspectors concluded it impacted the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) more than the attributes of the other Cornerstones. The inspectors utilized IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding screened as very low safety significance (Green) since the inspectors answered No to each of the questions in Exhibit 2, Section A, Mitigating Systems Screening Questions. The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of Problem Identification and Resolution, Self-Assessment, and involving the organization routinely conducting self-critical and objective assessments of its programs and practices. Specifically, the failure to identify the overly restrictive definition of SCAQ during previous audits of the CAP was caused by an insufficiently self-critical audit focus.
05000263/FIN-2016004-0131 December 2016 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationWelding Blanket Partially Covered Reactor Building Ventilation Intake (CAP 1539781) The following violation of very-low significance (Green) was identified by the licensee and was a violation of NRC requirements and met the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation (NCV). The licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, on October 28, 2016, when the licensee failed to follow procedures while performing activities affecting quality. Specifically, the licensee failed to identify and control modifications of safety-related SSCs in accordance with FPEMOD03; Temporary Modifications, in that operators installed a welding blanket which partially blocked the suction of VAC10A and 10B (intakes of the Reactor Building main supply fans) to prevent welding sparks from being sucked into the intakes and failed to follow steps in that procedure. Procedure FPEMOD03, Revision 13 states, in part, that This procedure shall be applied to Safety-Related SSCs, should be applied to augmented quality or reliability related SSCs, and may be applied to commercial facility changes. Contrary to these requirements, the licensee failed to use FPEMOD03 to evaluate the physical change of installing welding blankets over Safety-Related Reactor Building Ventilation main supply fan intakes for potential plant impact prior to installation. Specifically, this resulted in an increase in negative pressure of the reactor building and an increase of steam chase temperatures which had the potential to upset plant stability by initiating a Group 1 Isolation. This was identified by the licensee during a deliberate observation process by the Shift Manager. Immediate corrective actions included stopping the welding, removing the welding blanket, reducing the steam chase temperature. The licensee documented this issue in the corrective action program (CAPs 1539781, 1541340, and 1541514). The performance deficiency was determined to be more than minor because it adversely affected the Configuration Control attribute of the Initiating Events Cornerstone, with the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as Green based on answering no to the Initiating Events screening questions in inspection manual chapter (IMC) 0609 Appendix A, The Significance Determination Process for Findings at Power, effective July 1, 2012. The issue was entered into the corrective action program as CAPs 1539781, 1541340, and 1541514. The inspectors concluded the issue was licensee-identified based on the guidance in IMC 0612, Power Reactor Inspection Reports, issue date May 06, 2016.
05000263/FIN-2016004-0231 December 2016 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationPast Reactor Building Wide Range Gas Monitor Settings Prevented Transition to Mid/High Range (CAP 1537833) The following violation of very-low significance (Green) was identified by the licensee and is a violation of U.S. Nuclear Regulatory Commission (NRC) requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation. Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) requires, in part, that a holder of a license under this part shall follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E, and the planning standards of Title 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4) requires standard emergency classification and action level scheme, the bases of which includes facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, from June 30, 1994, through September 1, 2016, the licensee failed to maintain the effectiveness of the sites emergency plan and the emergency classification and action level scheme. Specifically, the licensee changed the Engineering Unit Conversion Factor (EUCF) for the Reactor Building (RB) Vent Wide Range Gas Monitor (WRGM), resulting in non-conservative monitor indications that were 13 times lower than the actual effluent levels. The EUCF error impacted the licensees Emergency Plans effectiveness (emergency classification and action level scheme) by reducing the licensees ability to rely on the monitor to identify radiological conditions that exceed the Emergency Action Level (EAL) initiating condition threshold for the declaration of Emergency Classification Levels ranging from an Unusual Event (UE) up to and including a General Emergency (GE). The use of the inaccurate RB Vent WRGM readings would delay the classification of an UE (RA1.2) due to actual effluent levels exceeding the threshold initiating conditions for the respective EAL, while the WRGM was erroneously indicating a much lower value. While this condition would prevent using the RB Vent WRGM for the declaration of an Alert (RA1.2), Site Area Emergency (SAE) (RS1.1), or GE (RG1.1), the licensee remained capable of performing the timely and accurate declaration of an Alert, SAE or GE by monitoring radiological conditions (releases) using the Off-gas Stack WRGM, in accordance with the bases identified in the respective EALs. Consequently, the Alert, SAE, and GE declaration (based on radiological conditions) would not be delayed or missed due to the RB WRGM issue. The NRC determined that since the change in the EUCF would have only prevented the timely and accurate classification of a potential UE (as required by 10 CFR 50.47(b)(4)) the issue was determined to be of a very-low safety significance (Green) as indicated in Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 22, 2015. On September 1, 2016, this issue was identified through the licensees self-assessment process and documented in the CAP as Action Request 01533526, Reactor Building Vent Wide Range Gas Monitor Effluent Channel Reading Non-Conservative. The licensee implemented corrective actions to correct the EUCF for the RB Vent WRGM and restore compliance. As such, the NRC determined this to be a Non-Cited Violation in accordance with Section 2.3.2 of the Enforcement Policy.
05000263/FIN-2016003-0130 September 2016 23:59:59MonticelloSelf-revealingFailure to Follow Procedures While Performing Activities Affecting QualityInspectors identified a self-revealed finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of Technical Specification 5.4.1.a, on June 24, 2016, when the licensee failed to follow procedures while performing activities affecting quality. Specifically, the licensee failed to accomplish activities affecting quality in accordance with FPGDOC03; Procedure and Work Instruction Use and Adherence, in that operators performed the Standby Gas Treatment (SBGT) A Train, Quarterly Test (025301) and failed to follow steps in that procedure. This resulted in an unanticipated trip of the turbine building ventilation and reactor building exhaust plenum fans causing an increase of steam chase temperatures which had the potential to upset plant stability by initiating a Group 1 Isolation. Immediate corrective actions included restoring ventilation to reduce the steam chase temperature, and entering the issue into the licensees Corrective Action Program (CAP 1526310). The inspectors determined that the licensees failure to follow procedures while performing activities affecting quality was a performance deficiency requiring evaluation. The finding was determined to be more than minor because it adversely impacted the Initiating Events Cornerstone attribute of Human Performance in the area of human error, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to follow procedures resulted in conditions that had the likelihood to upset plant stability and challenge critical safety functions, in this case, the potential to initiate a Group 1 Isolation due to high steam chase temperatures. The inspectors evaluated the finding in accordance with IMC 0609 and determined it to be of very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency; Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools (H.12).
05000263/FIN-2016010-0130 September 2016 23:59:59MonticelloSelf-revealingFailure to Plan and Perform maintenance to Correct HPCI Oil LeakA self-revealing finding preliminarily determined to be of low to moderate safety significance (White), and an associated apparent violation of Technical Specification 5.4.1.a, were identified for the licensees failure to plan and perform maintenance affecting the safety-related high pressure coolant injection (HPCI) system in accordance with written documents appropriate to the circumstance as required by Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, improperly planned and performed pre-April 2005 maintenance initiated a crack in a safety-related HPCI oil pipe and, for numerous years, the licensee failed to perform maintenance to resolve repeated identification of HPCI oil leakage. These failures resulted in a sudden increase in oil leakage on March 22, 2016, extending the unavailability of HPCI during a maintenance window and causing a loss of safety function. The licensee documented the issue in the corrective action program (CAP) as CAP 1516361 prior to repairing the oil leak and restoring the HPCI safety function. The inspectors determined that the licensees failure to pre-plan and perform maintenance on safety-related equipment was a performance deficiency; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors determined the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, improperly planned and performed 2005 maintenance initiated a crack in a safety-related HPCI oil pipe and, for numerous years, the licensee failed to perform maintenance to resolve repeated identification of HPCI oil leakage. These failures resulted in a sudden increase in oil leakage on March 22, 2016, extending the unavailability of HPCI during a maintenance window and causing a loss of safety function. The inspectors applied IMC 0609, Attachment 4, and IMC 0609, Appendix A, Exhibit 2, Section A, for Mitigating Systems to screen this finding and determined a detailed risk evaluation was required because the finding represented a loss of system and/or function. Therefore, a coordinated effort between inspection staff and regional Senior Reactor Analysts (SRAs) was required to arrive at an appropriate risk evaluation for the degraded condition that resulted from the finding. The SRA used the Monticello Standardized Plant Analysis Risk (SPAR) model, version 8.24 for the detailed risk evaluation. This evaluation concluded that the HPCI system was degraded for over 10 years and significantly degraded for approximately 4 months. The system is risk-important and is used to mitigate many internal and external initiating events. The total delta CDF for the 121 day portion of the exposure period is 3.8E6/yr., which is a finding of low to moderate safety significance (White). HPCI is an important high pressure injection system that is used to mitigate internal events, internal flooding, and internal fire events at Monticello. The inspectors determined the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Conservative Bias because licensee individuals failed to use decision-making practices that emphasize prudent choices over those that are simply allowable (H.14). Specifically, licensee Operations and Engineering management did not ensure entry into formal evaluation processes to address a potentially degraded condition for the HPCI oil leaks.
05000263/FIN-2016011-0130 September 2016 23:59:59MonticelloSelf-revealingFailure to Plan and Perform maintenance to Correct HPCI Oil LeakTitle 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, requires in part, that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are corrected. Licensee procedure FP-PA-ARP-01, CAP Action Request Process, Revision 11, Section 1.2 and Revision 42, Section 2.1, required, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, between March 14, 2006 and March 21, 2016, the licensee failed to correct oil leakage from the safety-related HPCI system, a condition adverse to quality. Specifically, as documented in Condition Reports nos. 1018528; 1508130; and 1515945, the licensee initiated a number of work orders and subsequently closed them without any further work performed to correct these conditions adverse to quality, which resulted in gradual degradation and loss of HPCI system safety function. This violation is associated with a White Significance Determination Process finding.
05000263/FIN-2016403-0131 March 2016 23:59:59MonticelloLicensee-identifiedLicensee-Identified Violation
05000263/FIN-2016001-0131 March 2016 23:59:59MonticelloNRC identifiedFailure to Use Procedures While Performing Activities Affecting QualityAn NRC identified finding of very low safety significance (Green) and associated of 10 CFR 50, Appendix B, Criterion V; Instructions, Procedures, and Drawings, was identified on February 5, 2016, as a result of the licensees failure to use procedures while performing activities affecting quality. Specifically, the licensee failed to accomplish activities affecting quality in accordance with FP-G-DOC-03; Procedure and Work Instruction Use and Adherence, in that documented procedures were not used to install a conduit support on safety related Emergency Filtration Train (EFT) Division II conduits. Immediate corrective actions included removal of the support and entering the issue into the licensees Corrective Action Program (CAP) 1511349. The finding was determined to be more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the inspectors based this determination on the fact that performing activities affecting quality without using procedures has the potential to adversely affect the design/qualification of a Structure, System, and Component (SSC) or impact the operability or functionality of a system or component. The inspectors determined the finding to have very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, teamwork because of the licensees work group failures to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained.
05000263/FIN-2016008-0131 March 2016 23:59:59MonticelloNRC identifiedFailure to provide acceptable Alternate Methods of Decay Heat RemovalThe inspectors identified an Unresolved Item associated with Technical Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System Cold Shutdown. Specifically, the licensee failed to verify that the capability of the alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay heat load. The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual Heat Removal Shutdown Cooling System Cold Shutdown, required in Mode 4, two RHR shutdown cooling subsystems shall be operable, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation. The TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem consisted of one operable RHR pump, one heat exchanger, the associated piping and valves, and the necessary portions of the RHR Service Water System System capable of providing cooling water to the heat exchanger. The TS Bases Section 3.4.8 further indicated that the two subsystems have a common suction source and were allowed to have a common heat exchanger and common discharge piping. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be operable. Since the piping and heat exchangers were passive components that were assumed not to fail, they were allowed to be common to both subsystems. When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown cooling subsystems inoperable, the Required Action was to, verify an alternate method of decay heat removal was available for each inoperable RHR shutdown cooling subsystem. The completion time for the required action was 1 hour, and once per 24 hours thereafter. The TS Bases 3.4.8 for Condition A indicated that with one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem was capable of providing the required decay heat removal. However, the overall reliability was reduced, therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the re-establishment of backup decay heat removal capabilities, similar to the requirements of the LCO. The bases further stated that the required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Alternate methods that can be used included (but not limited to) the Reactor Water Cleanup System by itself or using feed and bleed in combination with Control Rod Drive System or Condensate/Feed Systems. Abnormal Procedure, Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling, provided instructions for establishing alternate methods for decay heat removal. The inspectors noticed that except for the alternate method as described below in the G-EK-1-45, the licensee was not able to show by calculation or demonstration that the systems and methods credited in this procedure would be capable of providing sufficient heat removal capability or appropriate levels of redundancy as required by TS 3.4.8. The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold Shutdown Capability Report, dated April 22, 1981. This letter provided a report which described the capability of the Monticello Nuclear Generating Plant to achieve cold shutdown using only safety class systems and assuming the worst single failure. The alternate shutdown decay heat removal method used in the report credited combinations of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR to ensure suppression pool water temperatures were below the design limit. This method utilized the core spray system and safety relief valves to circulate reactor inventory to remove decay heat from the reactor. The inspectors noted that calculations supporting the above alternate strategy utilized an RHR subsystem that could be inoperable and/or unavailable and therefore may not be credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while the plant was in mode 4, with a credited one subsystem inoperable, the licensees credited alternate decay heat removal method that relied on an RHR subsystem, to perform the required suppression pool cooling function. The inspectors were concerned that relying on the only operable RHR subsystem for the alternate method did not meet the intent of the TS requirement as described in the TS Bases. Furthermore, the inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed to verify by calculation or demonstrations that two additional redundant alternate decay heat removal methods existed with sufficient capacity to maintain the average reactor coolant temperature below 212 degrees Fahrenheit. During the inspection, the licensee indicated that the Boiling Reactor Owners Group was in the process of developing a draft TS Task Force Traveler to address the requirement of TS 3.4.8 and its Bases. Based on the information above, the inspectors were concerned that the plant Operations Manual was inadequate and failed to include alternate decay heat removal methods that would enable the licensee to comply with the requirement of TS 3.4.8. The Operations Manual was required per TS 5.4.1, Procedures, which required that written procedures shall be established, implemented, and maintained covering the emergency operating procedures. The inspectors determined that this issue was unresolved pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC review of these actions. The licensee entered the inspectors concerns into their Corrective Action Program as AR 01516098.
05000263/FIN-2015201-0131 December 2015 23:59:59MonticelloNRC identifiedSecurity
05000263/FIN-2015003-0230 September 2015 23:59:59MonticelloNRC identifiedFailure to Perform High Radiation Area Portable Fire Extinguisher SurveillancesThe inspectors identified a finding of very low safety significance and an NCV of Technical Specification (TS) 5.4.1.d when the licensee failed to implement procedures associated with Fire Protection Program Implementation, to ensure that required refueling outage surveillances were performed for fire extinguishers located in high radiation areas (HRAs). Specifically, between March 2007 and May 2015, the licensee failed to implement steps 9 and 10 of 1123, Portable Fire Extinguishers, which required weighing and verifying adequate hydrostatic testing of the fire extinguishers in HRAs on a refueling outage frequency. Corrective actions included surveillance process changes and evaluation of the current status of the high radiation area fire extinguishers which resulted in the determination that outside of the surveillance process, a separate work activity had exchanged all the affected extinguishers with ones that were current on their surveillances in May 2015. This issue was entered into the licensees Corrective Action Program (CAP) 1484257 The inspectors determined that the failure to implement HRA fire extinguisher surveillances was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Work Management aspect because of the failure to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority and the failure to identify the need for coordination with different groups or job activities
05000263/FIN-2015008-0230 September 2015 23:59:59MonticelloNRC identifiedInaccurate and Incomplete Information Documented on VT/PT Report FormsThe NRC staff identified an AV that is being processed through the traditional enforcement process because it appears to involve willfulness, impacts the regulatory process, and is associated with an ISFSI. The AV involves 10 CFR 72.11, Completeness and accuracy of information, which required, in part, that information required by Commission regulations be maintained by the licensee to be complete and accurate in all material respects. However, from approximately September 5 to October 17, 2013, the NRC determined that licensee contractors apparently willfully completed PT examination forms, a quality assurance record, with inaccurate developer dwell times. The NRC also determined that the licensee contractors apparently completed PT examination forms, a quality assurance record, with other inaccurate information. This information was determined to be material to the NRC because it had the potential to mislead the NRC and the licensee as to the suitability for service of the DSCs.
05000263/FIN-2015007-0230 September 2015 23:59:59MonticelloNRC identifiedFailure to Review for Suitability of Application of Safety-Related Relays Installed Beyond Their Service LifeThe inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure measures were established for the selection and review for suitability of application of materials, parts, equipment and processes that were essential to the safety-related functions of structures, systems and components. Specifically, the licensee failed to review for suitability of application of safety-related Agastat and General Electric relays that had exceeded their service life, a condition non-conforming to their design basis, to justify their continued service considering in-service deterioration. The licensee previously entered this finding into the CAP, and completed corrective actions to replace or evaluate some relays and implemented a program to address the remaining relays in a timely manner The finding was determined to be more than minor because, if left uncorrected, the issue had the potential to lead to a more significant safety concern. Specifically, these safety-related relays were installed in protective circuits such as reactor protection system, etc., and their failure could impact the proper operation of these protective schemes. The inspectors did not identify a cross-cutting aspect associated with this finding as it was not reflective of the licensees current performance.
05000263/FIN-2015003-0330 September 2015 23:59:59MonticelloNRC identifiedFailure to Identify Safe Shutdown Equipment Impacts in Fire Strategy ProceduresThe inspectors identified a finding of very low safety significance and an NCV of TS 5.4.1.d when the licensee failed to maintain procedures associated with Fire Protection Program Implementation, consistent with the Updated Safety Analysis Report (USAR), to ensure that fire strategy procedures accurately indicated safe shutdown (SSD) equipment. Specifically, on June 25, 2015, the licensee failed to maintain A.3-12-C, Condenser Room Fire Strategy, to ensure SSD equipment was appropriately identified. In this case, fire strategy A.3-12-C failed to identify any SSD equipment in the room, despite the fact that SSD cabling ran through the room and was included in the USAR Fire Hazards Analysis. Corrective actions included performance of an extent of condition review which identified 40 other fire strategies where safe shutdown cabling was not identified, and initiation of procedure changes to include the appropriate SSD equipment. This issue was entered into the licensees CAP (CAP 1484142). The inspectors determined that the failure to maintain fire strategy procedures to ensure that SSD equipment was identified was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Self-Assessment aspect because of the licensees failure to conduct self-critical and objective assessments of its programs and practices.
05000263/FIN-2015008-0330 September 2015 23:59:59MonticelloNRC identifiedFailure to Assess Contractor Control of QualityThe NRC staff identified an AV that is being processed through the traditional enforcement process because it appears to be associated with an ISFSI, which falls under traditional enforcement. The AV involves 10 CFR 72.154(c), Control of purchased material, equipment, and services which required, in part, that licensees assess the effectiveness of the control of quality by contractors and subcontractors at intervals consistent with the importance, complexity, and quantity of the product or services. However, from approximately September 5 to October 17, 2013, the NRC determined that the licensee apparently did not assess the effectiveness of the control of quality by contractors in that the licensee apparently did not monitor the work of contractors performing PT testing on DSCs Number 11 through #16.
05000263/FIN-2015003-0130 September 2015 23:59:59MonticelloNRC identifiedInadequate Evaluation of Refueling Floor Structural Steel BeamsThe inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, on September 3, 2008, licensee personnel failed to verify the adequacy of design when they failed to use correct section properties in their calculation of stresses on structural steel beams supporting the refueling floor for the increased spent fuel cask loading. Reevaluation of the beams using correct methodology resulted in the conclusion that the beams would not meet the design basis stress limits. Immediate corrective actions for this issue included initiation of a CAP, performance of a functionality assessment which concluded that the refueling floor remained functional but non-conforming, and creating compensatory measures which limited the refueling floor live load in the cask loading area (CAP 1492837). The inspectors determined that the licensees calculational methodology was contrary to the standard engineering principles applicable for determination of stresses in structural members, which resulted in a failure to meet Criterion III, Design Control, and was a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612 because it was associated with the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers (reactor building) protect the public from radionuclide releases caused by accidents or events. Additionally, More than Minor Example 3.j of IMC 0612, Appendix E, Examples of Minor Issues, was used to inform the more than minor screening. The inspectors used IMC 0609, SDP, Attachment 4, Initial Characterization of Findings, and Appendix A of IMC 0609 to screen this finding. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. As a result, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000263/FIN-2015003-0530 September 2015 23:59:59MonticelloNRC identifiedFailure to Provide Complete and Accurate Information in LER 05000263/2015-002-00The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9 due to the licensees failure to provide information to the NRC that was complete and accurate in all material respects in accordance with the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, on June 29, 2015, the licensee failed to include an accurate assessment of the safety consequences and implications of a loss of shutdown cooling event when they issued LER 05000263/2015-002-00. This LER included an inaccurate assessment of safety implications, stating that engineering calculations show a potential worst case maximum temperature of 115 degrees Fahrenheit (F). The inspectors identified that engineering models actually showed potential worst case temperatures of 25-26 degrees F higher, which could have challenged or exceeded fuel pool cooling design specifications. Corrective actions included issuance of a revision to LER 2015-002-00 which contained the correct engineering modeling results and associated discussion of safety implications. The licensee entered this issue into its CAP (CAP 1484633). This issue was of more than minor significance under the Traditional Enforcement Process because the NRC relies on licensees to identify and correctly report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., loss of shutdown cooling) was evaluated separately and determined to be a finding of very low safety significance as documented in the 2015 2nd Quarter Integrated Inspection Report (05000263/2015002-01). In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was of more than minor concern with relatively inappreciable potential safety significance and is related to a finding that was determined to be a more than minor issue. Consistent with Example 6.9.d.1, this represented an example where the licensee submitted inaccurate information in a required report, which resulted in expansion of the scope of the next regularly scheduled inspection and required LER revision. Because there was no finding evaluated with this violation, the inspectors did not assign a cross-cutting aspect to this issue.
05000263/FIN-2015003-0430 September 2015 23:59:59MonticelloNRC identifiedDrywell to Torus Vacuum Breaker Past OperabilityDuring the cycle preceding the 2015 refueling outage, two evaluations associated with torus to drywell vacuum breaker operation were developed due to issues identified in the first quarter 2014. These included: CAP 1417977, Failure of drywell-torus vacuum breaker to close, which identified an occasion of dual indication during Procedure 0143 procedure. A second occurrence was observed several days later and was documented in CAP 1418471, AO-2382A Torus-to-DW vacuum breaker closed indication anomaly. CAP 1420318, DW-Torus vacuum breaker work performed with inadequate PMT, identified the PMT following shaft sealing component (O-ring) replacement during the 2013 outage was not performed as planned. The licensee evaluations for these CAP conditions concluded the Drywell to Torus vacuum breakers were operable. However, neither evaluation specifically considered the effect of an interference between the vacuum breaker test lever and vacuum breaker test actuator stem. Since this specific mechanism was not addressed in these two evaluations, past operability of the torus to drywell vacuum breakers was questioned. As a result, the licensee established a past operability evaluation be conducted via CAPs 1479198 and 1478212. The licensee completed its past operability evaluation on June 26, 2015. After review, the inspectors conveyed a number of questions to the licensees engineering staff in regard to the past operability evaluation. Although the licensee provided responses for the majority of these questions during the remainder inspection quarter, the licensee had requested external input in regard to one of the inspectors questions. Specifically, inspectors questioned whether it was possible for the bottom of the lever arm to be at an elevation above the top of the actuator stem at valve disc full open and if so, could the valve test lever arm have come to rest on top of the actuator stem, potentially impacting the ability of the vacuum breaker valve to close. Upon the close of this inspection period, that input had not yet been finalized and made available to the inspectors. As a result, this issue was considered to be an unresolved item pending a review of the licensees response and past operability for CAPs 1479198 and 1478212, including and the licensee response to open inspector questions.
05000263/FIN-2015008-0430 September 2015 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 72.158, Control of special processes, requires, in part, that licensees establish measures to ensure that special processes, including welding... and nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. Contrary to the above, the licensee identified on May 10, 2014, the licensee failed to perform verifications of a calibrated leak test instrument used on DSC lid to shell welds in accordance with Procedure TN 61BT-61BTH-HSMLD, Helium Leak Testing for NUHOMS Systems, Revision 1. Procedure TN 61BT-61BTHHSMLD, Revision 1, performs helium leak tests to demonstrate compliance with TS 1.2.4.a, 61BTH DSC Helium Leak rate of Inner Seal Weld. Additionally, contrary to 10 CFR 72.158, on April 2, 2014, the licensee failed to ensure enough filler material was deposited to achieve the minimum depth of the shell to outer top cover plate weld on DSC 16 in accordance with Procedure 12751-MNGP-OPS, Spent Fuel Cask Welding: 61BT/BTH NUHOMS Canisters, Revision 0. Instructions for welding operations are provided in Procedure 12751 MNGP-OPS, Revision 0, to ensure in field fabrication is performed in accordance with the Final Safety Analysis Report design basis drawings. During a nuclear oversight review of 2013 dry cask storage loading operations, the licensee identified that the helium mass spectrometer leak detection, calibrated leak instrument verification stabilizations, were not performed in accordance with TN 61BT-61BTH-HSMLD, Revision 1. Specifically TN 61BT-61BTH-HSMLD, Revision 1, Steps 8.3 and 8.4, require the user to determine the final instrument indicated leakage rate with the calibrated standard open and closed. The procedure step requires the user to ensure the system stabilizes while determining these results. TN 61BT-61BTH-HSMLD, Revision 1, Note 2, defines a stable signal as no more than a 1.0 E-8 std cm3/sec deviation in the indicated signal in 60 seconds. The licensee determined that for DSC 11, 12, 14, 15 and 16, stabilization times were less than 60 seconds. Specifically for DSC 12, stabilization times with the calibrated standard open were performed in 24 seconds, and stabilization times with the calibrated standard closed were performed in 22 seconds.
05000263/FIN-2015007-0130 September 2015 23:59:59MonticelloNRC identifiedInadequate Quality Assurance Controls for Nitrogen Supply for the AN2 SystemThe inspectors identified a finding having very-low safety significance, and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to assure the nitrogen supply for the alternate nitrogen (AN2) system was controlled as safety-related in system specifications, drawings, procedures, and instructions. Specifically, the licensee did not confirm effective quality assurance controls were in place to ensure the bottled nitrogen was acceptable to support the safety-related functions of this system. The licensee entered this finding into the Corrective Action Program (CAP), and subsequently contacted the commercial nitrogen gas supplier to confirm that the vendors quality controls provided a sufficient basis to conclude that the AN2 system was operable. The finding was determined to be more than minor because if left uncorrected, the issue had the potential to lead to a more significant safety concern. Specifically, if the commercial (e.g., non-safety) gas supply vendor quality controls were not adequate to ensure contaminants such as moisture or particulates were excluded from the nitrogen gas bottles, it could potentially disable the AN2 systems capability to support manual operation of safety relief valves during post loss-of-coolant-accident mitigation. The inspectors did not identify a cross-cutting aspect associated with this finding as it did not reflect current performance.
05000263/FIN-2015008-0130 September 2015 23:59:59MonticelloNRC identifiedFailure to Perform Penetrant Tests in Accordance with Procedural RequirementsThe NRC staff identified an AV that is being processed through the traditional enforcement process because it appears to involve willfulness and is associated with an Independent Spent Fuel Storage Installation (ISFSI). The AV involves Title 10 of the Code of Federal Regulations (CFR) 72.158, Control of Special Processes, which required, in part, that the licensee establish measures to ensure that special processes, including nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. The licensee established TriVis Procedure 12751 QP-9.202, Color Contrast Liquid Penetrant Examination Using the Solvent-Removable Method, Revision 1, as the qualified procedure for use in Dry Shielded Canister (DSC) NDE PT. However, from approximately September 5 to October 17, 2013, the NRC determined that licensee contractors apparently willfully failed to follow the TriVis procedure for developer dwell times, while performing PT on 66 of 66 DSC closure welds examined. The NRC also determined that the licensee contractors apparently failed to follow other parts of the TriVis procedure.
05000263/FIN-2015002-0630 June 2015 23:59:59MonticelloSelf-revealingLoss of Electrical Buses and Shutdown Cooling (SDC) Due to Inadequate Procedure AdherenceA self-revealed finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified due to the failure to properly implement Procedure 0304-01, Safeguard Bus Loss of Voltage Protection Relay Unit Calibration Safeguards Bus No. 15. Specifically, electrical maintenance workers failed to comply with Step 20 which directed the installation of a jumper between terminals ZX10 and ZX11 in an electrical panel, when they incorrectly installed the electrical jumper between terminals ZX11 and ZX12. This resulted in the loss of the Division I safety related 4160 Volts Alternating Current (Vac), 480 Vac, and 125 Volts Direct Current (Vdc) electrical buses, which subsequently led to the loss of shutdown cooling (SDC) for approximately 3 hours and 15 minutes. Initial corrective actions for this issue included immediately invoking strict plant status controls to focus efforts on recovery, restoring the electrical buses and SDC to operation, and reinforcing risk recognition and human performance tools. This issue was entered into the licensees CAP (CAP 1477351) and a root cause evaluation (RCE) was in progress at the time this inspection period concluded. The inspectors determined that the issue was more than minor because it adversely impacted the Initiating Events Cornerstone attribute of Human Performance and Configuration Control, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors utilized IMC 0609, Appendix G for shutdown operations and determined that the issue was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency aspect because of the failure of licensee individuals to implement error reduction tools and the failure of the organization to plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12).
05000263/FIN-2015002-0130 June 2015 23:59:59MonticelloNRC identifiedFailure to Maintain Portable Fire Extinguishers in Accordance with Fire StrategyThe inspectors identified a finding of very low safety significance and an NCV of TS 5.4.1.d when the licensee failed to implement procedures associated with Fire Protection Program Implementation to ensure that portable fire extinguishers were maintained in accordance with the fire strategy. Specifically, on May 1, 2015, the licensee failed to implement fire protection p an procedures when they failed to control three portable fire extinguishers in the condenser room, a room housing safe shutdown cabling, in accordance with Fire Strategy A.3-12-C. In this case, inspectors found that of the four dry chemical extinguishers required to be stationed in the condenser room, two indicated that they were partially depleted and needed to be recharged, and a third extinguisher was missing entirely. Immediate corrective actions included recharging the partially depleted extinguishers and procuring a portable extinguisher to replace the missing one. This issue was entered into the licensees CAP (CAP 1477246). The inspectors determined that the failure to implement the fire strategy procedure to ensure that condenser room portable fire extinguishers were maintained was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612 Appendix B because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Because the plant was shut down, the inspectors assessed the significance of this finding in accordance with IMC 0609, Appendix G, the Shutdown Operations SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Identification aspect because of the failure to implement a CAP with a low threshold for identifying issues, and failure to ensure that individuals identify issues completely, accur tely, and in a timely manner in accordance with the program (P.1)
05000263/FIN-2015002-0230 June 2015 23:59:59MonticelloNRC identifiedFailure to Measure Interpass TemperatureThe inspectors identified a Green NCV of Title 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, for a failure to measure the interpass temperature while performing welding on diesel generator fuel oil modification supports. Consequently, welding was performed without the Code and Procedure required interpass temperature being Monitored on a number of welds, a parameter which can affect the mechanical properties of the material being welded. To restore compliance, the welder proceeded to measure the interpass temperatures on the balance of the welds and verified that the interpass temperature did not exceed that allowed by procedure. The licensee entered this issue into its CAP (CAP 1475767). The inspectors determined that this issue was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because the inspectors answered yes to the more than minor question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, absent NRC intervention, the welder would have completed all of the welds without having measured the interpass temperature, a welding parameter which can affect the mechanical properties (e.g., impact properties) of some materials being welded, and if left uncorrected could lead to a potential failure of the weld in service. In accordance with Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness, of IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, the inspectors checked the box under the Mitigating Systems Cornerstone because leakage on the Emergency Diesel Generator (EDG) fuel oil system could cause core decay heat removal to be degraded. The inspectors determined this finding was of very-low safety significance (Green) based on answering yes to the question in Part A of Exhibit 2, Mitigating Systems Sc reening Questions, in IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued on June 19, 2012. Specifically, the inspectors answered yes to the screening question If the finding is a deficiency affecting the design or qualification of a mitigating Structure, System, or Component (SSC), does the SSC maintain its operability or functionality? The welder proceeded to measure the interpass temperatures on the balance of the welds and verified that the interpass temperature did not exceed that allowed by procedure, and the issue did not result in the actual loss of the operability or functionality of a safety system. The inspectors determined that the primary cause of the failure to monitor the interpass temperature procedure was related to the cross-cutting component of Problem Identification and Resolution, Operating Experience (P.5). Specifically, the organization failed to effectively implement external operating experience in a timely manner.
05000263/FIN-2015002-0330 June 2015 23:59:59MonticelloNRC identifiedFailure to Maintain Secondary Containment and Standby Gas Treatment System Operable During OPDRV ActivitiesThe inspectors identified a finding of very low safety significance and an associated NCV of TS 3.6.4.1, Secondary Containment and TS 3.6.4.3, Standby Gas Treatment System (SBGT) because the licensee did not maintain secondary containment and the SBGT system operable as required during activities considered OPDRVs. Specifically, on April 14, 2015, and again on May 13, 2015, the licensee failed to classify activities associated with draining reactor inventory as OPDRVs while relying on an automatic isolation function for the drain path, and as a result failed to maintain required equipment operable during these activities. Once questioned by the inspectors, the licensee took action to control other outage related draining activities as OPDRVs and placed this issue into its CAP (CAP 1479284). The inspectors determined that the failure to maintain secondary containment and SBGT operable while an OPDRV was in progress was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, RCS, and containment) protect the public from radionuclide releases caused by accidents or events because the secondary containment boundary and the SBGT were not maintained operable during an OPDRV activity. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations SDP since the reactor was shut down. The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 4 and Appendix H for containment integrity findings. Using Appendix H, the inspectors concluded the finding had very low safety significance (Green) because decay heat was low and containment was deinerted. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Documentation aspect because of the failure of the licensee to create and maintain complete, accurate and up-to-date documentation (H.7).
05000263/FIN-2015002-0530 June 2015 23:59:59MonticelloSelf-revealingInadequate Clearance Order Results in Unplanned OPDRVA self-revealed finding of very low safety significance and an associated NCV of technical specification (TS) 5.4.1, Procedures, was identified on May 16, 2015, when the licensee failed to implement procedure FP-OP-TAG-01, Fleet Tagging, for equipment control activities associated with the Scram Discharge Volume (SDV). Specifically, the licensee failed to ensure that clearance order checklist 58972-03 restored valve I-CRD-R-26, an SDV instrument vent valve, to its normal position prior to returning the SDV system to service. As a result, during subsequent reactor coolant system (RCS) pressure boundary testing, RCS water leaked out onto the reactor building floor through the open vent line, creating an unplanned operation with a potential for draining the reactor vessel (OPDRV). This issue was entered into the licensees CAP (CAP 1479307). Immediate corrective actions included termination of the leakage by closing and capping the SDV vent line and resetting the scram. The site initiated an apparent cause evaluation (ACE), which was in progress at the end of the inspection period. The inspectors determined that the failure to adequately restore the SDV system to service in accordance with fleet tagging requirements was a performance deficiency requiring evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, because it adversely impacted the Initiating Events Cornerstone attributes of Configuration Control and Procedure Quality, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations significance determination process (SDP) since the reactor was in Mode 4 (cold shutdown). The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 2 for Initiating Events. Using IMC 0609 Appendix G, Attachment 3, for a Phase 2 analysis, the inspectors determined it to have very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Challenge the Unknown aspect because of the failure of individuals to stop when faced with uncertain conditions and the failure to ensure that risks are evaluated and managed before proceeding (H.11).
05000263/FIN-2015002-0730 June 2015 23:59:59MonticelloNRC identifiedOperations with a Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment OperableA violation involving a failure to have secondary containment operable during Operations with the Potential to Drain the Reactor Vessel (OPDRV) was identified. Specifically, from April 23, 2015 through May 8, 2015, Monticello Nuclear Generating Plant performed a total of three activities within two work windows without setting secondary containment, which is a violation of Technical Specification (TS) 3.6.4.1. The NRC issued Enforcement Guide Memorandum (EGM) 11-003, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel, Revision 2, on December 13, 2013, allowing for the exercise of enforcement discretion for such OPDRV-related TS violations, when certain criteria are met. The NRC concluded that Monticello Nuclear Generating Plant met these criteria during the activities for which the EGM was invoked. Therefore, I have been authorized, after consultation with the Director, Office of Enforcement, and the Regional Administrator, to exercise enforcement discretion and refrain from issuing enforcement for the violation. Between April 23, 2015 and May 1, 2015 and again between May 2, 2015 and May 8, 2015, the Monticello Nuclear Generating Plant (MNGP) performed OPDRV activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the RPV water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV activities. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued EGM 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliance with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a LOI event, (b) maintaining level monitoring to ensure secondary containment can be closed before inventory is drained to the RPV flange, (c) maintaining the capability to isolate the potential leakage paths, (d) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (e) prohibiting movement of recently irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this licensee event report (LER) for potential performance deficiencies and/or violations of regulatory requirements. The inspectors also reviewed the stations implementation of the EGM during the OPDRVs for which the EGM was invoked. Based on review of the following items, the inspectors determined that the licensee met the EGM requirements for discretion: 1. The inspectors observed that the OPDRV activities were logged in the control room narrative logs and that the log entry appropriately recorded that the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least 21 feet and 11 inches over the top of the RPV flange as required by TS 3.9.6. The inspectors also verified that at least one safety-related pump was available as the standby source of makeup designated in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the RPV flange was greater than 24 hours. 3. The inspectors reviewed Engineering Change documents which calculated the time to drain down during these activities and the feasibility of pre-planned actions the station would take to isolate potential leakage paths during these periods of time. 4. The inspectors verified that the OPDRVs were not conducted in Mode 4 and that the licensee did not move recently irradiated fuel during the OPDRVs. The inspectors noted that MNGP had in place a contingency plan for isolating the potential leakage path and verified that two independent means of measuring RPV water level were available for identifying the onset of LOI events. TS 3.6.4.1 required, in part, that secondary containment shall be operable during OPDRV. TS 3.6.4.1, Condition C, required the licensee to initiate action to suspend OPDRV immediately when secondary containment is inoperable. Contrary to the above, between April 23, 2015 and May 1, 2015 and again between May 2, 2015 and May 8, 2015, MNGP performed OPDRV activities while in Mode 5 without an operable secondary containment. Specifically, the station performed the following OPDRV activities without an operable secondary containment: 12 Recirculation System pump upper seal replacement; 12 Recirculation System modifications to add and replace valves; and 11 Recirculation System modifications to add and replace valves. Because the violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation (EA-15-130). In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 12 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The inspectors observed that Monticello is tracking the need to submit a license amendment request in its CAP (CAP 1476012). LER 05000263/2015-001-00 is now closed. This event follow-up review constituted one sample as defined in IP 71153-05.
05000263/FIN-2015002-0430 June 2015 23:59:59MonticelloNRC identifiedFailure to Fill the Reactor Cavity in Accordance with Refueling Preparation ProcedureThe inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, on April 15, 2015, when the licensee failed to implement procedure 9001, Reactor Well & Dryer-Separator Storage Pool Filling Procedure, for refueling preparation activities. Specifically, when faced with indications that the condensate storage tanks (CSTs) did not contain enough water inventory to complete outage critical path reactor pressure vessel (RPV) flooding activities, the licensee failed to implement 9001 procedure steps for using prescribed equipment and methods to fill the reactor cavity. With the proceduralized methods unavailable, operators used the site decision-making process to utilize demineralizer water hoses to fill the cavity rather than processing required 9001 procedure changes. This issue was entered into the licensees CAP (CAP 1474891). Immediate corrective actions included action to initiate the procedure change process for 9001 and department communication to Operations regarding the incident, emphasizing that the decision making process is not a substitute for the procedure change process. The inspectors determined that the failure to fill the reactor cavity in accordance with the 9001 reactor well filling procedure was a performance deficiency requiring evaluation. The inspectors evaluated IMC 0612, Appendix E, and did not find any similar examples of minor issues. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the operations crews use of the decision-making process to support outage critical path by bypassing proceduralized steps and performing activities using methods contrary to the procedure could lead to a more significant safety concern. In addition, if performed incorrectly (i.e. without flushing the hoses prior to use), the use of demineralizer hoses could introduce foreign material into the core and challenge the integrity of the fuel cladding barrier. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations SDP since the reactor was in Mode 5 (refueling). The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 4 for Barrier Integrity and determined to have very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Conservative Bias aspect because of the failure of the individuals to use decision-making practices that emphasize prudent choices over those that are simply allowable, and the failure to ensure that proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop (H.14).
05000263/FIN-2015001-0931 March 2015 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.47(b)(14) and 10 CFR Part 50, Appendix E, Section IV.F.1. In part, Title 10 CFR 50.47(b)(14) states, Periodic exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Additionally, Title 10 CFR Part 50, Appendix E, Section IV.F.1 states, The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. The Monticello Emergency Plan, Section 8.1.2.4, describes the required demonstration periodicity for drill and exercises. Contrary to the above, on January 1, 2015, the licensee failed to perform four emergency preparedness drill objectives at the required frequency listed in the Monticello Emergency Plan, Section 8.1.2.4. Specifically, Objectives 11.01, 11.03, and 11.04 were required to be performed annually and were not performed in 2014. Additionally, Objective 11.04 was required to be performed semi-annually and was only performed once in 2014. All missed objectives were associated with radiological exposure controls. The NRC determined that the failure to comply with the established drill and exercise program was a degradation of a planning standard function in accordance with 10 CFR 50.47(b)(14) and was a very low safety significance issue (Green) as indicated in IMC 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. The licensee entered this issue in the corrective action program (CAP 1463920). As such, the NRC determined this to be an NCV in accordance with Section 2.3.2 of the Enforcement Policy.
05000263/FIN-2015001-0431 March 2015 23:59:59MonticelloNRC identifiedInadequate Evaluation of Operating Crew During Simulator AssessmentThe inspectors identified an URI on March 16, 2015, due to the licensees potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. In accordance with IMC 0612, Power Reactor Inspection Reports, the inspectors determined that this issue represented an URI because more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. On March 16, 2015, the NRC inspectors observed a potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. Specifically, during an NRC observation of a Licensed Operator Training self-assessment and emergency preparedness objective demonstration, the inspector observed that the evaluators may not have adequately critiqued a knowledge deficiency in the Interpreting and Diagnosing Events competency area when evaluating a Shift Managers (SM) performance. The Shift Managers performance could have adversely impacted EAL classification during a graded self-assessment. This assessment included an evaluated Drill/Exercise Performance (DEP) opportunity for the EAL classification in question. During the inspectors observation, they noted that the critique session did not appear to adequately probe why the classification-related performance weaknesses occurred, and did not appear to determine a course of specific actions for the crew to take to improve individual performance relative to the SMs role in the EAL classification. Specifically, the inspectors noted that at the end of the critique, this item was not discussed as an item needing resolution, nor was it discussed that the SM had a challenge to his qualifications and needed potential remediation, which appeared to be contrary to the sites MTCP0349 procedure. These discussions and follow-up actions did not take place until after the critique had concluded and the NRC inspectors raised questions about the SMs misinterpretation of Safety Parameters Display System (SPDS) and his overall performance. This item represents an issue of concern about which more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. The NRC inspectors will work to obtain additional guidance and clarification/interpretation of the existing guidance in order to resolve this issue. Corrective actions for this issue included disqualifying the individual, developing a remediation plan, and initiating procedure changes to improve the critique process. This issue was entered into the corrective action program as CAP 1470975. (URI 05000263/201500104, Inadequate Evaluation of Operating Crew During Simulator Assessment)
05000263/FIN-2015001-0531 March 2015 23:59:59MonticelloSelf-revealingTwo Emergency Diesels Inoperable Due to Human ErrorA self-revealing finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified on December 28, 2014, due to the failure to properly implement Procedure 0187-02B, 12 Emergency Diesel Generator /12 ESW (Emergency Service Water) Monthly Pump and Valve Tests. Specifically, operations personnel failed to comply with Step 42 which directed the 12 EDG local governor control switch to be lowered to idle setting. The failure to implement the actions directed by Step 42 resulted in the 11 EDG being inoperable. Corrective actions for this issue included procedure revisions to require: protection/flagging of redundant equipment when technical specification equipment is declared inoperable for any reason, including planned maintenance and surveillance; peer checking or concurrent verification for manipulation of operable technical specification related equipment; and all equipment manipulations require a hard match (between procedure and equipment labeling). This issue was entered into the licensees corrective action program (CAP 1460675). The issue was more than minor because if left uncorrected, the failure to properly implement procedures associated with safety-related equipment would have the potential to lead to a more significant safety concern. Specifically, the failure to follow procedure resulted in the 11 EDG being made inoperable coincident with the 12 EDG being inoperable. The inspectors utilized IMC 0609 and determined that the issue was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency aspect because of a failure of individuals to implement error reduction tools (H.12).
05000263/FIN-2015001-0331 March 2015 23:59:59MonticelloNRC identifiedFailure to Maintain a Standard Emergency Action Level Scheme for FloodingThe inspectors identified a finding of very low safety significance and an NCV of Title 10 CFR 50.54(q)(2) and 10 CFR 50.47(b)(4) for the licensees failure to maintain the effectiveness of the emergency plan. Specifically, from May 28, 2014, until February 26, 2015, the HA1.6 Emergency Action Level (EAL) threshold was in conflict with the EAL basis for the alert classification. Additionally, both the revised EAL threshold and original NRC-approved safety evaluation report EAL threshold were later found to be greater than the actual river level that could lead to damage of safe shutdown equipment. The licensees corrective actions documented that the current river level was 906 and if flooding were to occur the licensee would have relied on Procedure A.6, "Acts of Nature," and that an event response team would have been formed to monitor river level during the duration of a flood event. The licensee concluded that the shift manager, Event Response team, and plant management would have monitored for indication of degraded performance of equipment or structures necessary for safe shutdown for event classification escalation to the Alert level. The licensee entered this issue into the Corrective Action Program (CAP 1454593). The inspectors determined that establishing a flooding EAL threshold that was in conflict with approved EAL basis as required by 10 CFR 50.47(b)(4), and subsequent failure to determine the actual level that could lead to damage of safe shutdown equipment for the alert classification High River Level EAL HA1.6 was a performance deficiency. The inspectors determined that the issue was more than minor because it is associated with the Procedure Quality attribute of the Emergency Preparedness (EP) cornerstone and adversely affected the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Evaluation aspect because the licensee did not thoroughly evaluate the identified engineering error issue to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000263/FIN-2015001-0631 March 2015 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.1, Offsite Dose Calculation Manual, (ODCM) which requires in part, that licensee initiated changes to the ODCM shall be effective after approval of the plant manager. Contrary to the above, ODCM01.01 Revision 6 and ODCM02.01 Revision 10, were not approved by the plant manager prior to implementation. This was identified by the licensee as part of the self-assessment process. The licensee documented this issue in the corrective action program (CAPs 1455999 and 1462092). This finding was determined to be of very-low safety significance (Green) because it was not a failure to implement an effluent program and public dose did not exceed Appendix I of 10 CFR 20.1301(e) criteria.
05000263/FIN-2015001-0731 March 2015 23:59:59MonticelloLicensee-identifiedLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.11 which requires in part, that the Primary Containment Leakage Rate Testing (LRT) Program shall be in accordance with the guidelines contained in RG 1.163, Performance-based Containment Leak-Test Program, dated September, 1995. RG 1.163 directs use of ANSI/ANS56.81994, Containment System Leakage Testing Requirements as an acceptable testing standard. ANSI/ANS56.81994 states, in part that for pressure decay testing, temperature shall be recorded at the start and end of each test, and the leakage rate shall be calculated using a specific formula which incorporates this temperature data to temperature-compensate the volume lost. Contrary to these requirements, the licensees Containment Leakage Rate Testing Program failed to include direction to take temperature data and perform temperature compensation, which resulted in a failure to perform testing in accordance with the ANSI standard and RG 1.163. Specifically, during this time, the licensee failed to correctly perform pressure decay testing for approximately 44 containment penetrations, including the Personnel Airlock. Upon discovery, engineers performed a bounding engineering analysis which verified the containment barrier remained operable but nonconforming and entered the issue into the corrective action program (CAPs 1463917 and 1465869). The performance deficiency was more than minor because the issue is associated with the barrier performance reliability attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that the physical containment barrier protects the public from radionuclide releases. Specifically, the repeated failure to ensure containment leakage testing met technical specification and regulatory requirements was programmatic, affected multiple components, adversely affected LRT test accuracy, and consequently impacted the licensees ability to verify the containment barrier remained operable. The finding was of very low safety significance because the finding did not represent an actual open pathway in the physical integrity of the containment barrier and did not result in a loss of containment barrier operability. (Green)